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Title: Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

Abstract

The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

Inventors:
; ;
Publication Date:
Research Org.:
Westinghouse Electric Corp
OSTI Identifier:
7201345
Patent Number(s):
US 5309487; A
Application Number:
PPN: US 7-903683
Assignee:
Westinghouse Electric Corp., Pittsburgh, PA (United States) OAK; EDB-94-100102
DOE Contract Number:
AC03-90SF18495
Resource Type:
Patent
Resource Relation:
Patent File Date: 24 Jun 1992
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; PWR TYPE REACTORS; REACTOR ACCIDENTS; STEAM GENERATORS; MITIGATION; RUPTURES; AUXILIARY WATER SYSTEMS; ENGINEERED SAFETY SYSTEMS; PRIMARY COOLANT CIRCUITS; ACCIDENTS; AUXILIARY SYSTEMS; BOILERS; COOLING SYSTEMS; ENERGY SYSTEMS; ENRICHED URANIUM REACTORS; FAILURES; POWER REACTORS; REACTOR COMPONENTS; REACTOR COOLING SYSTEMS; REACTORS; THERMAL REACTORS; VAPOR GENERATORS; WATER COOLED REACTORS; WATER MODERATED REACTORS; 220900* - Nuclear Reactor Technology- Reactor Safety; 210200 - Power Reactors, Nonbreeding, Light-Water Moderated, Nonboiling Water Cooled

Citation Formats

McDermott, D.J., Schrader, K.J., and Schulz, T.L.. Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems. United States: N. p., 1994. Web.
McDermott, D.J., Schrader, K.J., & Schulz, T.L.. Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems. United States.
McDermott, D.J., Schrader, K.J., and Schulz, T.L.. Tue . "Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems". United States. doi:.
@article{osti_7201345,
title = {Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems},
author = {McDermott, D.J. and Schrader, K.J. and Schulz, T.L.},
abstractNote = {The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Tue May 03 00:00:00 EDT 1994},
month = {Tue May 03 00:00:00 EDT 1994}
}
  • The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolatemore » the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.« less
  • Results from an experimental investigation of steam generator tube rupture in the Semiscale Mod-2B system are presented. From the experimental results, the characteristic system response signature for a wide range of number of tubes ruptured has been described. The tube rupture was assumed to occur during normal full power operation (15.6 MPa system pressure, 37/sup 0/K core differential temperature). In addition, recovery scenarios involving operator actions were examined. The recovery scenarios included use of pressurizer auxiliary spray and internal heaters, steam generator feed and steam, primary feed and bleed, and main cooling pump operation. Recovery scenarios suggested by typical USmore » pressurized water reactor emergency operating procedures were followed.« less
  • The DPRA-SGTR computer program was written to develop a dynamic event tree for the analysis of a steam generator (SG) tube rupture (SGTR) event. Using the dynamic event tree, a full-scope understanding of the possible responses of a plant following an SGTR event and the related actions with the emergency operating procedures (EOPs) can be analyzed. RELAP5/MOD3.2 was linked to DPRA-SGTR to calculate the thermal-hydraulic response of a Westinghouse three-loop pressurized water reactor at the Maanshan nuclear power plant. One SG tube with a double-ended break was postulated at the beginning of the accident. The plant thermal-hydraulic behaviors, status ofmore » the mitigation systems, and operator actions following the EOPs were explicitly modeled in the postulated SGTR. A total of 131 sequences were generated after an SGTR event. Among the 131 sequences, 91 sequences with a frequency sum of 8.5 x 10{sup -6} were stopped either because of low-occurrence frequency (<1 x 10{sup -12}) or because the preset mission time was reached (30 000 s after initiating the event). Seven out of the 91 sequences with a frequency sum of 6 x 10{sup -9} were intentionally stopped as a fatal error occurred when RELAP5 was calculating the thermal-hydraulic response.« less
  • To reduce the temperature difference between a straight tube bundle and the housing surrounding the same in a steam generator assembly for pressurized water reactors, a preheater for feed water is provided, and part of the pressurized water, after it has flowed through the heat exchanger or steam generator proper, is used for heating the feedwater in the preheater. 3 claims, 1 drawing figure.